Journal of Engineering Advancements Vol. 01(02) 2020, pp 53-60 https://doi.org/10.38032/jea.2020.02.004
*Corresponding Author Email Address: [email protected] Published by: SciEn Publishing Group
Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with
and without Emergency Core Cooling System
Pronob Deb Nath*, Kazi Mostafijur Rahman, Md. Abdullah Al Bari
Department of Mechanical Engineering, Khulna University of Engineering & Technology, Khulna-9203, BANGLADESH
Received: June 05, 2020, Revised: June 26, 2020, Accepted: June 26, 2020, Available Online: June 29, 2020
ABSTRACT
This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of
Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe
in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit
line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion
and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried
out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic
model-based software developed using real data and computational approach incorporating reactor physics and control system was
employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different
operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small
break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant
dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been
investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow
through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the
operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core
meltdown and release of radioactivity after a specific time.
Keywords: LOCA; VVER-1200; PCTRAN; ECCS; Thermal-hydraulic behavior.
This work is licensed under a Creative Commons Attribution-NonCommercial 4.0 International License.
1. Introduction
VVER-1200 is the utmost type of VVER pressurizer water
reactor (PWR) meaning Vodo Vodyanoi Energetichesky
Reaktor or water-water energetic reactor with improved safety
features design and developed by OKB Gidropress, Russia. It is
a thermal neutron pressurizer reactor used water not only as
coolant but also as moderator. It consists of two circuit namely
the primary circuit, where the primary coolant receives heat from
reactor core and the secondary circuit where the heat supplied to
produce electricity [1] with four cooling loops, main circulation
pump, accumulator tanks of the ECCS, relief and safety
emergency valves. It’s said to be a generation III+ reactor with
improved safety features like sacrificial concrete core catcher
layer, passive heat removal system, etc. Inside the core, the
primary coolant deposit at 155 bar so that it can avoid
vaporization at high temperature [2]. Any leakage in the coolant
circuit can lose large amount of coolant from Reactor Cooling
System (RCS) result in Design Basis Accident or Beyond Design
Basis Accident according to the flow from RCS.
.
Design-basis accident (DBA) is a postulated accident that a
nuclear facility must be designed and built to withstand without
loss to the systems, structure and components necessary to ensure
public health and safety. DBA concept is very feasible for
abnormal operations in a nuclear power plant to vindicate
performance requirements for reactor system, structure and
components after anticipated operational occurrence. Beyond
Design Basis Accident (BDBA) which have extremely low
probability of happening is a sequence of accident which may or
may not lead to core degradations. DBA postulate for each type
of reactor covering different type of failure combinations like
reactivity-initiated accident, loss of flow transients, Loss of
coolant accidents, fuel handling accidents and so on. The DBA
is prescribed to different limiting condition for cladding
temperature, local fuel cladding oxidation, ZrO2 mass cladding,
whole body dose and others [3].
LOCA is one of most important and common DBAs for
each reactor design depending on size of break. LOCA assumed
that one of inlet or outlet pipe from circulating pump to reactor
vessel completely broken or partially broken and free discharge
of primary coolant from both broken ends occur. LOCA can be
happening for different reasons like, material defects of different
systems, material fatigue in pipe/shell, external impacts like
heavy loads or missiles, device failure during operation due to
different reasons, inadvertent opening on primary system
boundary and so on. The nuclear reactor core has different
emergency system to mitigate different type of accident depend
on the incident or accident type.
There are various safety features in VVER 1200 reactor
plants. VVER 1200 reactors are capable of sustaining initiating
events as well as their consequence under accident conditions
during a long period within the boundaries of design safety
criteria and assured by following structural and design features
P.D. Nath et al. /JEA Vol. 01(02) 2020, pp 53-60
54
like (a) increased coolant volume above the core (b) increased
coolant volume in the primary circuit in respect to fuel mass and
thermal power of the core (c) increased capacity of pressurizer
(d) reliable natural circulation (e) considerable water inventory
in horizontal SGs on the secondary side [2].
The primary coolant system of reactor core is the system
which is heated at reactor core flows into steam generator. The
primary coolant flow through heat exchanger tube boiling the
coolant in secondary cooling system which flow outside of heat
exchanger tube that drive the turbine generator and creates
mechanical energy from heat energy. After that, the primary
coolant comes from steam generator flow back to reactor core
using compressed pump.
The leakage size within reactor coolant pressure boundary
divided into two group namely Large Break Loss of Coolant
Accident (LBLOCA) and Small Break Loss of Coolant Accident
(SBLOCA). Reduction of water temperature during Emergency
core cooling system injection drop in about 50℃-70℃ the break
can be categorized as small breaks and for large break reduction
of temperature about 120℃-150℃ [4]. The primary system
remains at high pressure after the occurrence of LOCA and the
core remain covered. Coolant flows inside the primary system as
soon as the pumps are tripped due to automatically or manually
gravity control phase separation occurrence. The next
subsequent events depend on the overall behavior of the primary
and secondary system. Automatic and operator-initiated
mitigation measurements influenced the overall behavior.
Reactor design, break size, location of the break, core bypass
system, operator interactions have influenced the scenario. Each
alternative action has much impact to make a huge difference on
the LOCA. The LBLOCA classical DBA for reactor core where
one of the inlet pipes from the circulating pump is completely
non-functional or free discharge of primary coolant from both
broken ends which referred as Double Ended Guillotine break.
LOCA have those following phase (a)BLOWDOWN PHASE,
0-20 seconds (b) BYPASS PHASE, 20-30 seconds (c) REFILL
PHASE, 30-40 seconds (d)REFLOODING PHASE, 40-250
seconds (e) LONG TERM COOLING >250 seconds.
The VVER-1200 has following active and passive system to
mitigate different accident: (i) The Primary Circuit Emergency
with Planned Cooldown and Spent Fuel Cooling (ii) Low
Pressure Emergency Injection System to supply Boric Acid
Solution (iii) Emergency Core Cooling System (iv) Passive Core
Flooding System (v) Emergency Boron Injection System (vi)
Emergency Gas Removal System (vii) Primary Overpressure
Protection (viii) Secondary Overpressure Protection (ix) Passive
Heat Removal System (x) Steam Generator Emergency
Cooldown System (xi) Main Steam-line Isolation System (xii)
Double Envelope Containment and Core Catcher [1].
Personal Computer Transient Analyzer (PCTRAN) is a
nuclear reactor transient and accident simulation tool that
operates on a personal computer for learning different nuclear
reactor behavior under different condition. Concepts of neutron
multiplication, criticality, thermal hydraulic safety parameters,
transient analysis, accidental dose estimation and Xenon
poisoning and feedback in live simulation statistical analyses can
be carried out. The PCTRAN-VVER1200 simulator version
1.2.0 available from Micro-Simulation Technology for VVER-
1200 reactor for thermal output of 3200 MWt with two loops of
reactor coolant system represented as loop A and loop B use for
the numerical simulation purpose. It has shown the major
systems and components of the reactor like RCS, the Emergency
Core Cooling System (ECCS), the Pressurizer with its
components, the Low-Pressure Injection System (LPIS),
Feed/bleed water system, Steam Generator, Accumulator,
different safety valves and different components of the NPP. The
components like pump, valves indicating red colored are
operating and open system and white indicating colored indicate
idle and closed in the mimic interactive. Those components can
be switch on or off just by clicking on them in the interactive
display. It has two interface, the main interface for different
component and Dose mimic interface for radioactivity display.
Transient behavior simulation has been successfully
benchmarked by using PCTRAN approved by International
Atomic Energy Agency (IAEA). PCTRAN based on end-users
input data with a reactor point kinetics model for analysis nuclear
safety related concepts and phenomena that could be easily
explained with the simulators along with understanding of
valuable technological differences in various designs of
PCTRAN products. A high-resolution color mimic of the
Nuclear Steam Supply System (NSSS) and containment displays
are the status of important parameters and allows simulation of
operator actions by interactive control.
Though the main purpose of this simulator is operator
training and a dynamic test to validate the control logics in
reactor regulating system (RRS), this simulator is also used for
research purposes. Like, Mollah review about PCTRAN
simulator for safety and transient analysis of PWR [5]. Ibrahim
et al. showed Simulation of Safety and Transient Analysis of a
Pressurized Water Reactor using the PCTRAN [6]. Yi-Hsiang
Cheng et al. integrated development of accident dose
consequences simulation software for nuclear emergency
response applications using PCTRAN software [7]. Also further
they analyze the Improvement of accident dose consequences for
nuclear emergency response applications using PCTRAN
simulator [8]. Khan and Islam published an article on a PCTRAN
based investigation on the effect of inadvertent control rod
withdrawal on the thermal-hydraulic parameters of a VVER-
1200 nuclear power reactor [9]. Fyza et al. analyzed the thermal-
hydraulic parameters of VVER-1200 due to loss of coolant
accident concurrent with loss of offsite power with PCTRAN and
conclude with PSAR (Preliminary Safety Assessment Report)
regarding LOCA [10]. Tube rupture in steam generator and
transient analysis of VVER-1200 using PCTRAN simulation is
conducted by Saha et al. [11].
Only a limited research works on VVER-1200 new
generation III+ reactor related topic has been conducted. In
this study, the thermal hydraulics behavior of VVER-1200
reactor due to LOCA with 500% area with and without ECCS
are simulated by using PCTRAN software. Thermal hydraulic
parameters like, coolant dynamics, heat transfer, reactor
pressure, critical heat flux, temperature distribution in different
sections of reactor core have also been investigated in the
simulation. The flow in the reactor cooling system, steam
generators steam with feed-water flow, coolant steam flow
through leak level of water in different section, power
distribution in core and turbine were plotted to analyze their
behavior during the operations.
P.D. Nath et al. /JEA Vol. 01(02) 2020, pp 53-60
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2. Methodology
The numerical analysis was carried out by PCTRAN
VVER1200 v-1.2.0. By this software package one can simulate
a variety of accident and transient condition for VVER-1200
reactor nuclear power plants. User can select initial conditions
corresponding to different parameter and initiate malfunctions
to analyze as desired. Following analysis had been conducted
by this software- normal operating control, stream line break or
LOCA, recirculation pump trip, turbine trip, station blackout,
inadvertent rod withdrawal or insertion, Boron dilution
transient, steam generator tube rupture, feedwater transient,
anticipated transient without scram and any combination of
above. The user can manually trip the reactor pump, open or
close valve, override ECCS, change operational set points. The
plant mimic system has the ability to freeze, back track, snap a
new initial condition, change the simulation speed, trend plot
by selected variable for conducting system analysis.
The simulator based on nuclear reactor theory and solution
technique by Six- group point kinetics model which is a point
kinetics equation with six delayed neutron groups and
reactivity control from external sources and feedback. The
model equations are described in the literature by A.S. Mollah
[5]. The initial steady state conditions of the plant were Reactor
Power at 100% with RC Pressure 155 bar and Core Average
Temperature 306.9ºC with BOC (Beginning of Cycle) of Steam
Generator (SG) pressure 70 Bar. After setting up the initial and
boundary conditions, the necessary operating parameters were
measured for the analysis of reactor performance. Default
values of VVER-1200 reactor were given which can be
changed to initial value as required. Different type of
malfunctions with initial value, fraction percentage, delay time
will be set initially to run simulation.
After completion of simulation, transient plot of different
variables like flow rate, coolant activity, concentration of
materials, dose rate, fraction, energy, level of water, mass of
different element, power of different system, pressure,
radioactivity, enthalpy temperature, void fraction could be
created with respect to time and other variables. The sequence
of simulation of the whole process is shown in Fig. 1.
Fig. 1 Steps/order of simulation in PCTRAN.
3. Computational Results and Discussion
In this computational analysis two cases have been
considered for investigation. First case accounted for a design
basis loss of coolant accident where the ECCS was in operation
and the second case was for a beyond design basis loss of
coolant accident without any type of ECCS and passive cooling
system. After conducting the computational analysis, the plot
relevant to different parameters were discussed.
3.1 Analysis of LOCA at cold leg with all ECCS
In this simulation 500 cm2 (25.23 cm diameter) break in
cold leg simulation have been conducted. LOCA with break at
cold leg has serious impact because safety injection system
makeup water lost through this break. The simulation carried
out using manually defined malfunction in the code control at
cold leg with 500 cm2 failure fractions after 5 seconds of the
simulator to run at steady state condition as shown in Fig. 2. To
minimize the coolant loss, the RCPs manually tripped. There
was a rupture icon in the cold leg with primary coolant loss rate
appears in the simulator. Safety injection systems started
automatically after reactor trip signal generated on because of
the low pressure.
Fig. 2 Graphical user interface of PCTRAN: Initiation of 500
cm2 cold leg LOCA.
Fig. 3 Graphical user interface of PCTRAN: Reactor trip and
actuation of HPIS, RB spray pump.
P.D. Nath et al. /JEA Vol. 01(02) 2020, pp 53-60
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(a) (b)
(c) (d)
(e)
Fig. 4 (a) Pressure of RCS, RB, SGs (bar) for cold leg break (b) Mass flow rate of RCS leak, reactor coolant loops, ECCS flow
water (kg/s) (c) Temperature of average fuel, RCS, peak clad, peak fuel, RB (℃) for cold leg break (d) Core thermal power (%) ,
Nuclear flux power (%), Turbine Load power (%) for cold leg break (e) Water level of core, pressurizer, RB sump, SGs (m) for
cold leg break.
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Fig. 3 shows that, ECCS actuated High Pressure Injection
System (HPIS) at 15s when the RCS pressure was lower than
the nitrogen pressure accumulator tank. Although HPIS along
was not enough to compensate primary coolant loss,
accumulator and Low-Pressure Injection System (LPIS) with
recovery refueling water storage tank, reactor building spray
help to provide boronated water. LPIS provide enough water to
maintain the core submerged and responsible for long term
cooling of residual heat in reactor. Compared to large break
LOCA, the depressurization occurs relatively slowly in
SBLOCA. Upon actuation of accumulator, ECCS water flow
started to exceed the coolant discharge rate and the reactor
vessel water level slowly rose again and core level water
restored as illustrated in Fig. 4 (e). The RCS pressure fall down
from initial steady 160 bar to about 3 bar after 200 second from
the onset of accident. The Steam Generator A (SG-A) and SG-
B pressure initially increased after accident from 74 bar to 81
bar in 30 second for sudden loss of coolant but after 65 second
both SGs reached to equivalent pressure 72 bar as shown in Fig.
4 (a). The process of depressurization was very rapid slowing
down the reactor coolant leak rate. After being emptied the
pressurizer during initial occurrence, RCS leak rapidly
decreased as the pressure altitude change fast. RCS leak
initially started after 5 seconds of the incident with 4150 kg/s
mass flow rate which slowly decreased with respect to time
after 40 seconds of the accident. The leak loss of mass flow
become equal with reactor coolant loop B flow at 105 seconds
which was about 390 kg/s. After 220 seconds the mass flow
rate become very slow but it needed more time to stop
completely. ECCS started after 85 seconds with gradual mass
flow rate highest of 1390 kg/s. After 245 seconds, the ECCS
slow down. The reactor coolant loop A has 4300 kg/s mass flow
rate, coolant loop B has highest 13000 kg/s mass flow rate
which decrease very rapidly and stable in 150 seconds to 225
seconds at 390 kg/s in Fig. 4 (b). For SBLOCA initiation
reactor coolant release very high flow rate through break,
which loss rate decrease due to HPIS actuation and
depressurization of primary system.
In the meantime, emergency accumulator water system
active to provide enough makeup water in RCS to maintain the
core submerged in water so that during the transient event, fuel
and cladding temperature never rise to critical value shown in
Fig. 4 (c). Highest peak clad temperature and peak fuel
temperature can’t rise to critical value to melt the core as the
control rods are fall down after 7s and huge makeup water from
ECCS, HPIS active due to this transient event. Fuel peak
temperature, clad peak temperature, average fuel temperature
falls from respectively 800°C, 415°C, 320°C to 150°C, 140°C,
110°C after 300s. The sump water level in reactor building
increase from 40°C to 100°C due to recirculation through
RWST with LPIS. The accident is assumed to happened when
the reactor operating at full 100% power. Due to the occurrence
of LOCA in cold break, the simulation shows reactor trip after
40s which indicate core thermal power and nuclear flux fall to
5% and decay heat power continuously extend for next few
cycles to cooled down to zero in Fig. 4 (d). The turbine load
initially dropped after LOCA which increase for following 20s
and again decrease to zero as turbine trip after the malfunction.
3.2 Analysis of LOCA at cold leg without ECCS or any passive
heat removal system
In VVER-1200, the emergency core cooling system is
consisting with not only HPIS, but also it has LPIS, passive heat
removal system, emergency accumulator water system, feed
water system, reactor building spray pumps and refueling water
storage tank with it to prevent all kind of accident including
DBA and BDBA to tackle the situation in emergencies.
Whenever ECCS face any kind of accident signal, it
automatically got actuated to keep the core under water and
cooled it down in order to eliminate any serious impact on the
total system and environment with working personnel. When
the ECCS can’t operate as needed, then this scenario can turn
into severe accident with core meltdown and fuel degradation.
In this case, all the ECCS equipment were disabled and
reactor coolant pump was turned off manually, with a 500 cm2
LOCA in cold leg. Though, it is a hypothetical cause with
unavailability of all those ECCS equipment but following the
Fukushima Daiichi accident caused by earthquake and tsunami
where all the emergency system failed and level/class 7 disaster
happened, possible consequences of this type of accident needs
to be analyzed.
For LOCA accident without enabling the ECCS, the
reactor trips instantaneously and after 9 seconds all the control
rods trip down to reactor core with pressurizer
emptied within 20 seconds due to very rapid depressurization
as shown in Fig. 5 (a). The fuel temperature continuously
increases due to absent of all ECCS and passive heat removal
system. With increased cladding temperature, the interaction
between the cladding and steam was accelerated. Hydrogen
concentration inside reactor increased with time and shortly
after the core uncovery, as depicted in Fig. 5 (b) about 980
seconds after the accident, the hydrogen concentration in the
reactor building exceeded 5%. Fuel temperature increased as
time goes by resulting the reactor core starts to melt. In Dose
Mimic, the core melting process can be observed in Fig. 5 (c).
Molten fuel collapse into the bottom of the vessel as there is
nothing that can cooled the core materials. The vessel lower
head may then heat up to the melting point, too. Fig. 5 (d)
illustrates that, the molten debris with molten core concrete
interaction with vessel penetration may drop into the
containment cavity floor takes around 3000 seconds for 500cm2
LOCA at cold leg without ECCS. During the fuel damage
process, first the fission gas in the cladding may leak out. Later,
if the fuel and cladding continue their degradation,
radionuclides will also be released. In addition to iodine and
noble gases, there are alkali metals, tellurium, barium, cerium,
lanthanides, etc. The elevated concentration of these
radionuclides would find their ways through the vessel break,
relief valves, and containment leakage into the environment.
The molten metal interacts with concrete and forms a slump. At
lower temperatures, degassing of concrete occurs and both
steam and carbon dioxide can be released. At higher
temperatures concrete can also be molten and mixed with
metals. In the event that the molten core heats up the vessel
bottom and melts through it, the debris, called corium, falls into
the reactor cavity.
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(a) (b)
(c) (d)
Fig. 5 Graphical user interface of PCTRAN: (a) Reactor trip for LOCA without ECCS (b) Complete core uncovery (c) Beginning
of core melt (d) Molten core-concrete interaction (MCCI) with vessel penetration.
For LOCA at cold break, the reactor coolant system leak
and loops fluid decrease rapidly. At the time of emptying of
pressurizer in first few seconds, coolant leak rate rapidly
decreased as the pressure elevation of break firstly change.
From Fig. 6 (a), level of reactor building sump slowly grew up
to a certain point until the recirculation process terminated and
there left no water to carry the poison. Level of core water
started to decrease very slowly at about 73 seconds. There
created a tube rupture loop at cold leg after 350 seconds. In Fig.
6 (b), different temperatures were plotted as respect to time for
initial 300 seconds. Due to increasing the sump in reactor
building, temperature also increased as followed by the sump.
Initially after the accident, the control rod fell down to the core
but without the activity of any coolant flow inside the core, the
power slowly rose by chain reaction. Before that, the core water
level and control rods minimized the temperature about 140
seconds until all the remaining water vaporized totally. After
the core fuel emersion, temperature rapidly increased due to
unavailability of coolant. Increase in Zircaloy cladding
temperature with fuel temperature accelerated the reaction
between Zirconium and water. Hydrogen gas accumulation
inside reactor building occurred quickly at the top of building
which can get exploded when it increased to 5%. The
Zirconium cladding distortion increased very high; fraction rate
about 42 in 1500 seconds whereas Hydrogen buildup was 8%
at that time. DNBR or departure from nuclear boiling is the
ratio of critical heat flux at specific location to operating local
heat flux at that location. In all NPP, DNBR must be greater
than one but it can’t allow to increase more than 95. From the
Fig. 6 (c), it is observed that the DNBR increased very slowly
as local heat flux can’t able to cross limit until all the coolant
system loss and its critical value was about 80. When the local
flux was higher than critical heat flux, the core materials melted
down for increasing heat flux. Fuel Doppler reactivity is the
fractional change in reactivity caused by fuel temperature. Until
it’s less than one, there was no harm of reactivity but after
explosion of the reactor building beyond 980 seconds, reactive
substance released to atmosphere. As it increased after the loss
of coolant accident, it progressed very much and spread
radioactive materials as shown in Fig. 6 (d). Fig. 6 (e) shows
the comparison between LOCA for different temperature in
fuel (TF), average fuel (TAVG), peak fuel (TFPK), peak clad
(TPCT) with and without ECCS for first 300s. Initially after the
accident, all ECCS activated when the control rod submerged
with fuel inside the core, hence temperature fell down for both
conditions. But unavailability of ECCS in second case,
temperature began to rise when all the water inside core
vaporized and fuel with control rod became unstable due to
active heat generation.
P.D. Nath et al. /JEA Vol. 01(02) 2020, pp 53-60
59
(a) (b)
(c) (d)
(e)
Fig. 6 (a) Level of core, pressurizer, SGs, RB sump water for LOCA at cold leg without ECCS (b) Temperature of average fuel,
RCS, peak clad, peak fuel, RB (℃) for LOCA cold leg break (c) DNBR for LOCA without ECCS (d) Fuel Doppler (%), rod (%),
modifier reactivity (%) for LOCA without ECCS (e) Temperature of different component with and without ECCS for LOCA at
cold leg.
P.D. Nath et al. /JEA Vol. 01(02) 2020, pp 53-60
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4. Conclusion
In this study, numerical study has been conducted to
investigate the different thermal hydraulics characteristics of
nuclear power plant due to loss of coolant accident with and
without ECCS. Loss of coolant accident may be a type of
postulate event but the probability of happening it in any kind
of nuclear reactor plant have enough chance due to different
reasons. LOCA accident with both design basis accident
analysis and beyond design basis accident analysis are
employed in this paper. From the results following conclusion
could be drawn:
Design Basis Accident and Beyond Design Basis Accident
both must be mitigated before they are harmful for
environment and mankind. A simple Design Basis
Accident can turn into serious disaster if it won’t handle
properly. Risk of radioactive material can be spared if
proper maintenance isn’t taken care off. Serious accident
is not only dangerous for human, but also for every life on
that area.
Different type of thermal hydraulics behaviors like
pressure, flow rate, water level in different sections, clad
fuel temperature, feed water cooling system, power and
many things can be investigated through LOCA analysis.
From numerical simulation for LOCA without ECCS
transient behaviors and core degradation dose mimic it is
found that it takes 3066 seconds to complete core
meltdown for 500% area or (25.23 cm diameter) failure
fracture. Different arbitrary failure fraction takes different
time to conclude in MCCI but reactivity release must be
controlled. Design basis LOCA show good agreements
with Primary Safety Analysis Safety report (PSAR) but in
case of BDB LOCA occurring without ECCS, the core
melting process is not able to mitigate the accident.
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